Purex process that has been conventionally used as a reprocessing technique for a light-water reactor fuel is known as a wet reprocessing process of spent nuclear fuels.
In this process, after shearing a spent nuclear fuel, small pieces of the fuel are dissolved in nitric acid [dissolution step]. The nitric acid solution includes uranium, plutonium and FP (fission products), and insoluble residues of FP and solid impurities such as cladding tube chips produced during shearing. Thus, these residues and impurities are removed, and the nitric acid concentration and the like are adjusted [clarification and adjustment step]. This nitric acid solution is brought into contact with a mixed organic solvent of n-dodecane and TBP (tributyl phosphate) to extract uranium and plutonium into the organic solvent phase [extraction step], and to subject FP left in the nitric acid aqueous phase to vitrification as a high level liquid waste. The organic solvent phase containing uranium and plutonium is brought into contact with a nitric acid aqueous phase containing uranium (IV) or HAN (hydroxylamine nitrate) to back extract plutonium into the nitric acid aqueous phase and leave uranium in the organic solvent phase [distribution step]. The organic solvent phase containing uranium is further brought into contact with a diluted nitric acid solution to back extract uranium into the nitric acid aqueous phase. The obtained nitric acid solution containing uranium and the obtained nitric acid solution containing plutonium are subjected respectively to extraction, washing, back extraction, and condensation to remove impurities such as FP [purification step]. The purified nitric acid solution containing uranium and the purified nitric acid solution containing plutonium are also denitrated to recover a uranium oxide and a plutonium oxide respectively [denitration step].
In addition, development of reprocessing techniques represented by RETF (Recycle Equipment Test Facility), advanced wet reprocessing processes, and advanced wet reprocessing processes adopting direct extraction have been proposed as improved Purex processes (Non-Patent Documents 1 and 2).
The development of reprocessing techniques for fast reactor fuels in RETF has started on the basis of Purex process which has been put to practical use as a reprocessing technique of a spent nuclear fuel from a light-water reactor. Since the performance of each step of dissolution, clarification, and extraction needs to be improved in response to increasing FP contents due to increased fast reactor fuel burnup, the equipments such as continuous dissolvers, centrifugal clarifiers, and centrifugal extractors have been newly developed. However, separation of uranium and plutonium from a nitric acid solution of a spent nuclear fuel is carried out in accordance with Purex process.
In the advanced wet reprocessing process, some of a large amount of uranium included in a nitric acid solution of a spent nuclear fuel is separated and recovered in advance by an uranium crystallization method in which the temperature dependence of the uranium solubility is used to precipitate and separate uranium, thereby reducing the amount of nuclear materials to be treated in a subsequent solvent extraction step and the following steps. In order to use as a MOX (mixed oxide) fuel, an uranium oxide and a plutonium oxide that are recovered in conventional Purex process, have been mixed at an appropriate ratio. However, the uranium/plutonium ratio in a nitric acid solution of a spent nuclear fuel to be subjected to Purex process is controlled in advance to an appropriate ratio suitable for a MOX fuel, and uranium and plutonium are extracted into an organic solvent and then back extracted all together from the organic solvent, thereby allowing only an uranium/plutonium mixed oxide of the appropriate ratio to be obtained. Such a process has been proposed in the advanced wet reprocessing process.
The advanced wet reprocessing processes adopting direct extraction has been proposed as one of alternative techniques to the advanced wet reprocessing process, where a solvent in which nitric acid and TBP form a complex (TBP nitric acid complex) is directly brought into contact with a solid spent nuclear fuel, thereby selectively recovering uranium and plutonium. This process can be simplified as compared with the advanced wet reprocessing process, and reduces nitric acid for dissolution and the amount of liquid waste generated from an extraction step, and thus the reduction of the step for concentrating a high level liquid waste can be thus expected.
The above-described wet reprocessing processes include a dissolution step for dissolving a spent nuclear fuel in nitric acid, and nitric acid in washing liquids produced in separation or distribution of and a purification process of uranium or plutonium is mostly recovered by a nitric acid recovery system such as concentration by evaporation and reused, while some of the nitric acid results in excess nitric acid. This excess nitric acid is neutralized with a sodium salt selected from sodium hydroxide, sodium hydrogencarbonate and sodium carbonate, thereby producing a sodium nitrate liquid waste. It is to be noted that the term “sodium salt” as used herein refers to sodium hydroxide, sodium hydrogencarbonate or sodium carbonate.
It is often the case that liquid wastes produced in the analyses carried out in each step of the wet reprocessing process are also nitrate forms although occurring in slight amounts, and if these liquid wastes are acidic wastes, the wastes are subjected to neutralization processing with a sodium salt, thereby producing a sodium nitrate liquid waste.
Further, an off-gas produced in dissolving the spent nuclear fuel in the nitric acid includes nitrogen oxides, a slight amount of radioactive components and the like. The off-gas is thus washed with an alkali solution of a sodium salt (alkali scrubbing) in order to remove the nitrogen oxides, the radioactive components and the like, thereby producing a sodium nitrate liquid waste as the washing liquid waste.
In addition to this, an organic solvent degraded due to radiation is washed with a sodium salt in order to remove the degraded component. The liquid waste of the sodium salt is neutralized with nitric acid to produce a sodium nitrate liquid waste.
These sodium nitrate liquid wastes produced as secondary wastes through the wet reprocessing process are subjected to the processing such as concentration by evaporation, and the condensate liquid is discharged. Since the concentrated liquid is a low-level radioactive waste, the radioactive waste is subjected to vitrification, cement solidification or bituminization, or kept liquid, powdered, or pelletized for intermediate storage.
On the other hand, various reduction methods have been conventionally proposed, which may be catalytic reduction methods in which nitrate-nitrogen in liquid waste water containing nitrate-nitrogen is reduced to nitrogen gas with a reducing agent and catalyst or supercritical reduction methods in which the nitrate-nitrogen is reduced to nitrogen gas with a reducing agent under supercritical conditions where water serves as a supercritical fluid.
Methods using hydrogen as the reducing agent and methods using a reducing agent other than hydrogen are known as the catalytic reduction methods. The methods using hydrogen as the reducing agent include, for example, a method in which nitrate-nitrogen is reduced to nitrogen with a zeolite catalyst in the presence of hydrogen (Non-Patent Document 3). The methods using a reducing agent other than hydrogen include, for example, a method in which a reducing agent such as hydrazine is added to waste water containing nitrate-nitrogen, and the waste water is brought into contact with a sponge copper catalyst to reduce the nitrate-nitrogen to nitrite-nitrogen, and then further brought into contact with a palladium catalyst with hydrazine or the like to reduce the nitrite-nitrogen to nitrogen gas (Patent Document 1).
The supercritical reduction methods include, for example, a method in which a nitrogen in a nitrate is reduced to nitrogen gas with a reducing agent such as an alcohol, ammonia, carbohydrate, formic acid, or oxalic acid under conditions where water serves as a supercritical fluid, that is, at a temperature and a pressure equal to or more than the critical point (374° C., 22 MPa) of water (Patent Document 2).
Non-Patent Document 1: “Development on FBR Fuel Reprocessing Technology”, Japan Society of Mechanical Engineers (No. 96-3), 5th Power and Energy Technology Symposium '96, Collected Papers for Lectures
Non-Patent Document 2: “Component Technologies Development of Reprocessing System—Advanced Aqueous Reprocessing Process Technologies Development—”, Report from Japan Nuclear Cycle Development Institute, No. 24 Separate Volume, 153-164, November 2004
Non-Patent Document 3: “Development of Wet Reductive Decomposition Processing Technique for Nitrate-Nitrogen in Wastewater”, Report on Completion of Program for Promotion of Technology Development from 2002 to 2004, I-26, RITE-Wakamatsu Second Laboratory, Research Institute of Innovative Technology for the Earth, March 2005
Patent Document 1: Japanese Patent Laid-Open No. 2003-126872
Patent Document 2: Japanese Patent Laid-Open No. 2005-241531
As described above, the sodium nitrate liquid wastes produced as secondary wastes through the wet reprocessing process of a spent nuclear fuel are subjected to processing such as concentration by evaporation, and the condensate liquid is discharged. Since the concentrated liquid is a low-level radioactive waste, the radioactive waste is subjected to vitrification, cement solidification or bituminization for storage, or kept liquid, powdered, or pelletized for intermediate storage.
However, when a spent nuclear fuel is subjected to wet reprocessing, sodium nitrate liquid waste that are secondary waste are inevitably produced, and the amount of radioactive waste to be processed correspondingly increases. This issue becomes a big problem when a wet reprocessing process is carried out.
It is thus an object of the present invention to provide a novel and improved method that is capable of substantially reducing the amount of radioactive waste generated due to a sodium nitrate liquid waste by reductively decomposing a sodium nitrate liquid waste produced through a wet reprocessing process of a spent nuclear fuel and recovering and reusing a sodium salt as a decomposition product.